The problem of control of the vertical instability is studied for a massless filamentary plasma with finite resistivity included for the shell and active control coil. Stability boundaries are determined. The system can be stabilized up to a critical decay index, which is predominantly a function of the geometry of the passive stabilizing shell. A second, smaller, critical index, which is a function of the geometry of the control coils, determines the limit of stability in the absence of derivative gain in the control circuit. The system is also studied numerically in order to incorporate the non-linear effects of power supply dynamics. The power supply bandwidth requirement is determined by the open-loop growth rate of the instability. The system is studied for a number of control coil options which are available on the DIII-D tokamak. It is found that many of the coils will not provide adequate stabilization and that the use of inboard coils is advantageous in stabilizing the system up to the critical index. A hybrid control system which utilizes such inboard coils on a time-scale which is faster than the vessel L/R time is proposed. Experiments carried out on DIII-D confirm the appropriateness of the model. Using the results of the model study, DIII-D plasmas with decay indices exceeding 90% of the critical index have been stabilized. Measurement of the plasma vertical position is also discussed.

The ITER plasma control system has the same functional scope as the control systems in present tokamaks. These are plasma operation scenario sequencing, plasma basic control (magnetic and kinetic), plasma advanced control (control of RWMs, NTMs, ELMs, error fields, etc) and plasma fast shutdown. This chapter considers only plasma initiation and plasma basic control. This chapter describes the progress achieved in these areas in the tokamak experiments since the ITER Physics Basis (1999 Nucl. Fusion 39 2577) was written and the results of assessment of ITER to provide the plasma initiation and basic control. This assessment was done for the present ITER design (15 MA machine) at a more detailed level than it was done for the ITER design 1998 (21 MA machine) described in the ITER Physics Basis (1999 Nucl. Fusion 39 2577). The experiments on plasma initiation performed in DIII-D and JT-60U, as well as the theoretical studies performed for ITER, have demonstrated that, within specified assumptions on the plasma confinement and the impurity influx, ITER can produce plasma initiation in a low toroidal electric field (0.3 V m −1), if it is assisted by about 2 MW of ECRF heating. The plasma basic control includes control of the plasma current, position and shape-the plasma magnetic control, as well as control of other plasma global parameters or their profiles-the plasma performance control. The magnetic control is based on more reliable and simpler models of the control objects than those available at present for the plasma kinetic control. Moreover the real time diagnostics used for the magnetic control in many cases are more precise than those used for the kinetic control. Because of these reasons, the plasma magnetic control was developed for modern tokamaks and assessed for ITER better than the kinetic control. However, significant progress has been achieved in the plasma performance control during the last few years. Although the physics basis of plasma operation and control is similar in ITER and present tokamaks, there is a principal qualitative difference. To minimize its cost, ITER has been designed with small margins in many plasma and engineering parameters. These small margins result in a significantly narrower operational space compared with present tokamaks. Furthermore, ITER operation is expensive and component damage resulting from purely operational errors might lead to a high and avoidable repair cost. These factors make it judicious to use validated plasma diagnostics and employ simulators to 'pre-test' the combined ITER operation and control systems. Understanding of how to do this type of pre-test validation is now developed in present day experiments. This research push should provide us with fully functional simulators before the first ITER operation.

Experiments on the TCV tokamak have shown that rapid vertical movement of diverted ELMy H-mode plasmas can affect the time sequence of Edge Localised Modes. The effect is attributed to the induction of an edge current during the movement of the plasma column in the spatially inhomogeneous vacuum field of a single null configuration. In TCV the fast vertical movement is provoked by the positional control coils inside the vacuum vessel, however it is argued that a similar effect might be produced in larger devices only using poloidal field coils external to the vessel. A simple model, which includes plausible elements of the dynamical behaviour of the edge pressure gradient and edge current, which together dictate the MHD stability of the discharge against edge-localised, current-driven modes, is used to reproduce some of the features seen in these experiments.

During the first year of operation, the TCV tokamak has produced a large variety of plasma shapes and magnetic configurations, with 1 . O B J1.46T, I <800kA, ~S2.05, -0.7G%0.7. A new shape control algorithm, Eased on a finite element reconstruction of the plasma current in real time, has been implemented. Vertical growth rates of 800 sec-', corresponding to a stability margin f=l.IS, have been stabilized. Ohmic H-modes, with energy confinement times reaching 8 h s , normalized beta (p ,aB/I> of 1.9 and z P R 8 9 -P of 2.4 have been obtained in singlenuB X-point deuterium discharges with the ion grad B drift towards the X-point. Limiter H-modes with maximum line averaged electron densities of 1 . 7~1 0~~m -~ have been observed in D-shaped plasmas with 360kASIp&00kA.

The ITER Integrated Modelling & Analysis Suite (IMAS) will support both plasma operation and research activities on the ITER tokamak experiment. The IMAS will be accessible to all ITER members as a key tool for the scientific exploitation of ITER. The backbone of the IMAS infrastructure is a standardized, machine-generic data model that represents simulated and experimental data with identical structures. The other outcomes of the IMAS design and prototyping phase are a set of tools to access data and design integrated modelling workflows, as well as first plasma simulators workflows and components implemented with various degrees of modularity.